B210059, Rev 1 to GE-NE-B2100594-1, Radiological Analyses of Design Basis Accidents
| ML20248L405 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 03/09/1998 |
| From: | Careway H, Pappone D, Tran P GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20248L403 | List: |
| References | |
| GE-NE-B2100594, GE-NE-B2100594-1-R01, GE-NE-B2100594-1-R1, NUDOCS 9803200176 | |
| Download: ML20248L405 (48) | |
Text
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NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT GE Nuclear Energy GE-NE-B2100594-1 175 Curtner Ave DRF B21-00594 San Jose, CA 95125 March 1998 RADIOLOGICAL ANALYSES OF DESIGN BASIS ACCIDENTS I
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TASK 24 i
Rev Prepared Date Reviewed By Date Approved Date Approved By Date i' By By NSP O- S.S. Wang 9-96 D.R. Rogers 9-96 P.T.Tran 9-96 D.M. Musolf I l-8-96 I D.C. Pappone H.A.Careway 3 P.T.Tran O )4 M v. J /9/95 i W& 3l% N O d~~~J %g $1rren 3l6/qg
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IMPORTANTINFORMATION REGARDING CONTENTS OF THIS REPORT :
1 PLEASE kEAD CAREFULLY The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the Power Rerate contract between Northern States Power Company (NSP) and GE,'as identified in Purchase Order Number PH0303SQ, dated -
December 27,1994, as amended to the date of transmittal of this document, and nothing
- contained in this document shall be construed as changing the contract. The use of this information by anyone other than NSP, or for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, GE makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy or usefulness of the information contained in this document,' or that its use may not infringe privately owned rights.
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I GENE.B21005941 TABLE OF CONTENTS l
- 1. GOAL OF RERATE PROGRAM 4
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- 2. TOPICAL EVALUATION BACKGROUND 5
- 3. EVALUATIONS 6 3.1 Control Rod-Drop Accident 6-3.1.1 Assumptions 6 3.1.2 Source Term for Offsite Dose Evaluations 7 3.1.3 Offsite Dose Results 7 3.2 Loss-of-Coolant Accident 13 3.2.1 Assumptions 13 3.2.2 Radiological Consequences Evaluation 14 3.3 Main-Steamline-Break Accident 21 3.3.1 Assumptions 21 3.3.2 Noble Gas Concentration 22 3.3.3 lodine Concentration' 22 3.3.4 Offsite Dose and Control Room Dose Evaluations 22 3.4 Fuel Handling Accident 30 3.4.1 Assumptions 30 3.4.2 Results of Calculation 31
- 4. REVIEW OF NRC COMMENTS 35
- 5. REVIEW OF GENERIC COMMUNICATIONS 36 5.1 10CFR100 37 5.2 Regulatory Guide 1.3 37 5.3 Regulatory Guide 1.5 37 i
- 6. CONCLUSIONS 37 j 4
7,
SUMMARY
OF RESULTS 38
- 8. REQUIRED ACTIONS 39 l
- 9. REFERENCES 45 3 '
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3 4 o .l.: GOAL OF RERATE PROGRAM
- L This analysis is being nerformed as an assessment in support of the power rerate .
P project. .The effect 'of an increase in reactor thermal power up to' 1880 MWT.
L . (112.6%) will be evalaated for each system to determine if this increase can'be L,, accomplished safely within existing system configurations.' The license submittal-l-4 . . will request approval to operate at 1775 MWT (106.3%).- The objective of the f report will be to determine if the. system is capable' of performing its design-S function at the: increased power level, to' determine if any' modifications am required to support the power increase and to evaluate plant reliability. ;The -
evaluation will identify the' differences in system operation for both new power .
levels;
. Increasing rated power provides the most cost effective use of existing equipment.
In most cases, there is ~ sufficient capability' in the _ equipment such that no
. significant decrease in margin is rpquired to operate at the higher level.
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- 2. TOPICAL EVALUATION BACKGROUND
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This task is being performed as an assessment in support of the Monticello Power Rerate Project. Radiological effects of limiting design basis accidents under increased reactor power up to 1880 MWt are evaluated. Design basis accidents
- which were analyzed -in : USAR Chapter 14 Plant- Safety. Analysis and are -
considered radiologically limiting under rerate operations ~ are: control-rod-drop-accident, loss-of-coolant accident, main steamline-break accident, and fuel-handling accident. Each accident is evaluated separately _in;the sections that
- follow. Offsite doses at the exclusion area boundary and the low population zone due to each accident are calculated as well as control room doses following loss-of-coolant accident and main-steamline-break accident. Control room doses following control-rod-drop accident- and fuel-handling accident were not calculated because they are bounded by the main-steamline-break accident dose.
To comply with Regulatory Guide 1.49 [1], all nccidents are evaluated under operating conditions corresponding to 102% of rerated power, or 1918 MWt. _ The only exception is the main-steamline-break accident, which is assumed to occur at hot standby conditions.
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p 3. EVALUATIONS 3.1 Control Rod-Drop Acc! dent The radiological evaluation of the control-rod-drop accident (CRDA) is described in Monticello USAR Section 14.7.1 [2].
3.1.1 Assumptions The postulated event is one in which a high worth control blade is stuck in the fully inserted position and is decoupled from its control rod drive. Sometime after, the control rod drive of this blade is withdrawn, and the control blade subsequently drops at the maximum speed and creates a localized ~ power 3 excursion.
In the fuel licensing document GESTAR 11 [3] it is assumed that due to the ,
power excursion, S50 rods in the 8x8 array fuel reach an enthalpy of 170 Cal /g, I which is the enthalpy limitation for cladding perforation. An 8x8 array is the standard configuration for GE9 and GE10 fuel assemblies. Reference 4 -
indicates that a similar power excursion in 9x9 array fuel results in 1000 rods reaching 170 Cal /g. Since the 8x8 array fuel typically consists of 60 or 62 fuel rods and 9x9 array fuel has 74 rods per bundle, damage assumptions using 8x8 q fuel with 60 fuel rods proves to be bounding, i Monticello utilizes banked position withdrawal sequence (BPWS) control. -1 GESTAR 11 confirms that for all plants with BPWS, the peak fuel enthalpy in a !
rod-drop-accident would not exceed the design limit of 280 cal /g, which assures that dispersal of fuel into the reactor coolant will not occur due to a rod-drop-accident. Compliance of GE12 fuel with the generic rod-drop-accident analysis j for BPWS plants is documented in Reference 5. Reference 5 confirmed the '
applicability of the bounding analysis for GE12 fuel by comparing the local
- peaking, Doppler coefficient, and rod worths of GE12 fuel with those used for the bounding analyses. It is concluded that the licensing limit 280 cal /g for rod-drop-accident analysis bounds all fuels up to GE12 (10x10 array) designs.
)
Standard Review Plan 15.4.9 [6] provides essential bases for the radiological ;
consequence analysis of a control-rod-drop acciden' In our analysis, it is 'l conservatively assumed.that about 1% of the damaged rods reaches fuel melt i temperature and releases 100% of the noble gas and 50% of the iodine rcxl i
- . ' inventory to the reactor coolant [6). For the cladding perforated rods, it is 4
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assumed that 10% of the noble gases and iodine fission products which were accumulated in the fuel-clad gap is released to the reactor coolant [7].
- Ninety percent of the iodine activities in the reactor coolant is retained by'the water.' The rest (10%) of the reactor coolant iodine activity and all of the noble
- gas activity are instantaneously transpcerted to the condenser, where 90% of the iodine plates-out in the condenser. All airborne activity in condenser is assumed to leak directly, without filtration, to the environment at the rate of 1% per day.
The release of activities from the condenser terminates in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (6].
3.1.2 Source Term for Offsite Dose Evaluations
~i' Table 3.1-l' summarizes the accident analysis parameters / assumptions for the radiological. evaluations of ~ control-rod-drop accident. The assumed core inventory at time of accident is the GE generic fission products source at time
. zero [8].E Major isotopes relevant to offsite dose evaluations are listed in Table 3.1-2.
The atmospheric dispersion factors to the site boundary exclusion area and low population zone were originally calculated by Bechtel [9]. They are included in Table 3.1-1 as part of the evaluation parameters.
3.1.3 Offsite Dose Results i
Calculated activities in the condenser air space and activities being released to the environment, as functions of time after the accident, are presented in Tables 3.1-3 1 and 3.1-4 respectively. Dose consequences at the exclusion area boundary and the low population zone from this accident'are presented in Table 3.1-5, together with the SRP 15.4.9 acceptance criteria, which is 1/4 of the 10CFR100 dose limits, i
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(- ~ GENE-B2100594-1 f-TABLE 3.1-1 ASSUMPTIONS FOR MONTICELLO RERATE CRDA ANALYSIS PARAMETER VALUES REF Power Level (MWt) 1918 [1]
Fuel Rod Damaged (8x8 fuel) 850 [2]
Cladding Failed rods 841 see text Melt rods 9-Power Peaking Factor for Damaged Rods 1.5 [6]
Activity Released to Coolant fiom cladding failed Rods (%)
Halogen 10 [7]
Noble Gas 10 Activity Released to Coolant from Melt Rods (%)
Halogen 50 [6]
Noble Gas 100 l
1 Coolant Halogen Reaches Condenser (%) -10 [6] l l
Condenser Halogen Airborne (%) 10 [6] ;
Condenser Leak Rate (%/ day) 1.0 [6] i Release Duration (hours) 24 [6]
Offsite Atmospheric Dispersion Factors (Sec/m') [9]
Condenser release (Ground level) l 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (EAB/LPZ) 9.20E-4/7.93E-5 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (LPZ) 7.93E-05 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> - 5.35E-05 1 - 4 days 2.28E-05 l 4 - 30 days 6.68E-06 l
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1 TABLE 3.1-2 CORE INVENTORY T=0 HR AFTER SHUTDOWN ISOTOPE Cl/MWt 1-131 2.63E+04 l132 3.85E+04 ,
1133 5.50E+04 l-134 6.06E+04 l135 5.20E+04 j KR-83m 3.14E+03 KR-85m. 6.73E+03 )(
KR-85 3.02E+02 KR-87 1.29E+04 ' I KR-88 1.83E+04
- j. + KR-89 2.28E+04 XE131m 1.5BE+02 XE133m 2.31E+03 XE 133 5.53E+04 XE135m 1.04E+04 XE-135 7.15E+03 XE-137 4.85E+04 XE-138 4.61E&O4 I
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l l I-TABLE 3.1-5 MONTICELLO RERATE CRDA OFFSITE DOSE (REM) 2-HOUR EAB 30-DAY LPZ Thyroid Whole Body Thyroid Whole Body Offsite Doses 0.92 .096 0.66 0.021 j SRP 15.4.9 Criteria (rem) 75 6 75 6 I
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1 3.2 Loss-of-Coolant Accklent ,
e The radiological evaluation of the loss-of-coolant accident (LOCA) is described in
' Monticello USAR Section 14.7.2 [2].
I 3.2.1' Assumptions Even though fuel failures we not predicted for the sequence of events in a LOCA, ,
,; the ' NRC ' guidance contained in ~ Regulatory Guide 1.3 [10]' states that the radiolorbal analysis should assume fuel melt and releases of 100% of noble gas fission products and 50% of halogen fission products in the core. Of the released ~
activity, half of the halogen plates.out in the containment building and half a ~
remains. airborne . in the . primary containment. The primary containment is
- assumed to leak at the maximum Technical Specification' leak rate of 1.2 (volume)
% per day [l1].
Leakage through the main steamline isolation valves (MSIVs) is included in the allowable <contaimnent leak rate. The Technical Specification maximum allowable MSIV leak rate is 46 scfh total for all four main steam lines.1 This was j calculated to be approximately 0.17%.of containment volume per day. To provide i
- additional conservatism'and account for other potential leakage pathways which
- bypass secondary containment, the MSIV leakage rate was increased to 0.3%/ day, 3
or 25% of the total allowable containment leakage in the analysis. This activity
, passes through steam lines into the main condenser and is released to the l environment at ground level. The balance (75%) of the containment leakage is
!' assumed to leak to the reactor building, where it is processed by the Standby Gas L , ' Treatment system (SGTS) before being dispersed from the plant stack. These y assumptions are consistent with the existing licensing basis analysis for rated j power [2]. j Non-organic iodine transported through steam lines and condenser is subject to plate-out and resuspension inside t'he pipes and the condenser. The BWR j
[ . Owners' Group (BWROG) Methodology for evaluating MSlV leakages and -
j
' condenser releases [12] is used in the analysis to calculate the effects of deposition and resuspension of airborne iodines. l I
The design basis flow rate of the SGTS is at least one air change per day [2].
However, in the analysis it is conservatively assumed that all the airborne activity in the reactor building is exhausted to the environment without delay. In other i words, no cn dit is given f ' radioactive decay due to building holdup nor activity
, dilution due to mixing. In the actual calculation, an arbitrarily large leak rate (106 13
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GENE B2100594-1 l I
%/ day) is used to simulate the instantaneous exhaust of activity from the reactor !
building. l A summary of parameters / assumptions used in the analysis is given below and listed in Table 3.2-1. i The reactor has been operating at the rated power for three years prior to the recirculation line break [13].
e 100% of the noble gas fission products and 50% ofiodine fission products are released from the core and become airborne as a result of the accident-[10).
- 50% of the airborne iodines platcout and condense in the drywell [10].
- The remaining airborne iodines are available for leakage from the primary I containment. 91% of these iodines are in elemental form,5% in particulate form, and 4% in organic iodide form [10].
- The effective suppression pool decontamination factor for iodines is 4.3 [14]. l e The primary containment is leaking at the maximum Technical Specification I permitted leak rate of 1.2% per day [11]. I e 25% of the primary containment leakage is assumed to bypass secondary ;
containment by leal:ing past the MSIVs where a dominant portion of the flow !
leaks into the condenser complex via the main steam drain lines [2].
- The effects of deposition, conversion, and resuspension for non-organic l fodines released via the MSIVs are taken into consideration [12].
e 75% of the primary containment leakage is to the secondary containment, where airborne activity is processed by the Standby-Gas-Treatment System (SGTS) and released immediately (without mixing with the secondary containment atmosphere) from the plant stack 12].
The SGTS starts function at time zero after the accident [17].
. The SGTS removes iodines at 85% filter efficiency to conservatively account for the effect of humidity in standby train cooling air [15].
The assumed core inventory at time of accident is the GE generic fission products source at time zero after shutdown [8]. Major isotopes of interest are listed in Table 3.2-2.
3.2.2 Radiological Consequences Evaluation v
3.2.2.1 Offsite Dose Results The atmospheric dispersion factors to the site boundary exclusion area and low population zone are included in Table 3.2-1. Calculated offsite doses at the exclusion area two hours after the accident and at the low population zone 30 days l
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! I after the accident are shown in Table 3.2-3, together with 10CFR100 offsite dose F limits.
3.2.2.2 Control Room Dose Evaluations The following control room parameters are applied in the control room dose evaluations: <
e Control room volume is 27,000 ft'[15].
. . Control room emergency filtration train (EFT) starts operating at time zero after the accident [2].
.
e' The intake flow rate of the EFT is 900 ft'/ min [17).
Infiltration flow bypassing the EFT is 250 ft'/ min until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident and 10 ft / min thereafter [16].
There is no recirculation feature in the control room EFT [2].
The atmospheric dispersion factors to the control room ventilation intake are included in Table 3.2-1. '
Calculated 30-day control room personnel thyroid dose and whole-body dose are presented in Table 3.2-3. Both are well within the GDC19 exposure limits.
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<;, GENEB2100594-1 TABLE 3.2-1 ASSUMPTIONS FOR MONTICELLO LOCA DOSE ANALYSIS PARAMETER VALUES BEF.
Power Level (MWt) 1918 [1]
fi Percent of Core Activity Airborne in Pri. Containment [10]
Noble Gas 100 lodine 25 fodine Forms (%) [10]
Organic 4 l
Elemental 91 Particulate 5 i Primary Containment Leakage Rate (*/dday) 1.2 [11]
Primary Containment Leak Bypassing Secondary 0.3 [2]
Containment (including MSIV leakage) (%/ day)
Pool Decontamination Factor (Non-organic lodines) 5 [14]
Main Condenser Leak Rate (/dday) 2.99 [12]
calculated Secondary Containment Drawdown Time (sec) 0 (17]
Secondary Containment Holdup None [17]
Effective SGTS Filter Efficiency (%) 85 [15]
16
--__-______-_______n
l 4 l GENE-B2100594-1 L
- i. l TABLE 3.2-1 (CONTINUED)
ASSUMPTIONS FOR MONTICELLO LOCA DOSE ANALYSIS PARAMETER VALUES REF Offsite Atmospheric Dispersion Factor (sec/m') (9)
Ground Level 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (EAB/LPZ) 9.20E-4/7.93E-5 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (LPZ) 7.93E-05 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 5.35E-05 1 - 4 days 2.28E-05 4 - 30 days 6.68E-06 I Stack Release 0 - 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (EAB/LPZ) 1.01 E-4/3.51 E-5 I
0.5 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (EAB/LPZ) 3.21 E-6/1.30E-6 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (LPZ) 1.30E-06 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 8.54E-07 1 - 4 days 3.70E-07 4 - 30 days 1.11 E-07 Control Room Intake Atmospheric [9] i Dispersion Factors (sec/m )
l Ground Level Release ;
O - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.67E-03 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.41 E-03 I 1 - 4 days 9.65E-04 l
4 - 30 days 5.62E-04 l Stack Release 1 - 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 2.98E-04 0.5 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 2.47E-11 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.55E-11 1 - 4 days 6.20E-12 4 - 30 days 1.66E-12 i
17
_.-_____._..________a
GENE-B2100594-1 TABLE 3.2-1 (CONTINUED)
ASSUMPTIONS FOR MONTICELLO LOCA DOSE ANALYSIS PARAMETER VALUES REF Data for Control Roorn Volume of Control Room (ft') 27,000 [15]
Filter intake (cfm) 900 [17]
Efficiency of Charcoal adsorber (%) 98 [16]
Unfiltered Inteakage (T < 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) (cfm) 250 [16]
Unfiltered Inteakage (T > 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) (cfm) 10 [16]
Recirculation Rate (cfm) 0 (2)
Occupancy Factor [18]
O - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.0 1 - 4 days 0.6 4 - 30 days 0,4 I8 1
GENE-B2100594-1 TABLE 3.2-2 CORE INVENTORY T=0 HR AFTER SHUTDOWN ISOTOPE Ci/MWt 1-131 2.63E+04 1
1-132 3.85E+04 l-133 5.50E+04 l-134 6.06E+04 l-135 S.20E+04 KR-83m 3.14E+03 KR-85m 6.73E+03 KR 85 3.02E+02 KR-87 1.29E+04 KR-88 -
1.83E+04 KR-89 2.28E+04 XE-131m 1.58E+02 XE 133m 2.31 E+03 XE 133 5.53E+04 XE-135m 1.04 E+04 XE-135 7.15E+03 XE-137 4.85E+04 XE-138 4.61 E+04 1
l 19 1
t
L <
s GENE-B21005941 J TABLE 3.2-3 1
MONTICELLO RERATE LOCA DOSE (REM) f CONTROL ROOM 2-HOUR EA 30-DAY LPZ Thyroid _Whole Bodv** Thyroid Whole Body Thyroid Whole Body i
13.4 .020 4.88 0.83 10.6 0.36
)
GDC19 Limits (rem) i 10CFR100 Limits (rem) 30 5 300 25 300 25 Does not include contributions of reactor building shine or cloud shines.
I 20
, 4
p ;,
s , ,
~
~
., . :1
\
L GENE.B2100594-t ,
y'
'3.3; Main-Steaanline-Break Accident q
~
, Thc1 radiological evaluation; of . main-steamline-break accident. (MSLBA) is . c
. ' described in Monticello USAR Section 14.7.3 [2]i 1 3.3.1 ' Assumptions The postulated accident invo.lves a guillotine break of one of the four main steam lines outside the containment, resulting in' mass loss from both ends of tlie break.
There is no fuel damage as a consequence of this event, therefore the only activity '
released to the environment is that associated .with the steam and liquid discharged -
from the break. Initially only steam will issue from the broken end of the steam line. Subsequently, rapid depressurization due to the break causes the reactor pressure vessel water level to rise, resulting in a steam-water mixture flowing .
~
from the break (blowdown)-until:the main steam. isolation valves are closed.
Consistent with USAR assumptions, in our analysis ~the MSIV closure time is
~
, assumed to be 10.5 seconds after the accident,1which is longer than that assumed
'in the HELB evaluation for EQ purposes [19]. - Activity associated:with the '
discharged coolant is airborne in the turbine building instantaneously and released - -;
to the environment without delay. !
q In our analysis, it is assumed that the accident occurs at hot' standby conditions.
-' At these conditions, steam generation from the decay heat in the core is very low and cannot make up the steam loss through the break. The results are high rate of -
, vessel depressurization and rapid rising of. water level to the main steamline inlet.
In addition to hot standby conditions, the. Appendix;K break flow model was assumed in order to maximize the two-phase break flow rate. ' Both of these >
H assumptions yielded the maximum coolant mass releases through the bmak [20]. 1 1
Four percent of current rated (or 3.5% of nrated) power, steam flow, and feedwater flow rate were used in the actual SAFER [20] calculations to generate coolant mass releases. Two cases are studied: the first case assumes reactor pressure initially is at the safety relief valve opening setpoint,"ll58 psia. The second case assumes the initial reactor pressure at the pressure regulator setpoint, l
- %5 psia. The result shows that total integrated mass leaving reactor pressure : j y1 vessel through the break is 86152 lb in the first case, of which 71574 lb is liquid. ;
In the second case it is 78617 lb,'of which 66223 lb is liquid [20]. !
Accident parameters relevant to the radiological analysis are summarized in Table !
g 3.3-1. The atmospheric disipersion factors to the site boundary and to the control j L; , room intake, as well'as control room parameters, are included in the same table. l
[ 4
, 21 f
GENE-B2100594-1.
3.3.2 ' Noble Gas Concentration -
The assumed noble gas activity is the Monticello Technical Specification limit
, which corresponds to an off-gas release rate of 0.26 Ci/sec (rounded off to 0.3 '
Ci/sec in the USAR) at 30 minutes delay. This activity is assumed to consist of a
- standard isotopic ' fraction- calculated in Reference 22, which was based on measurement data.
3.3.3 Iodine Concentrate.p_n The activity concentration of halogen isotopes in the primary coolant under .
normal operation,in units of pCi/ce,is given in Table 9.2-1 of the USAR. .In our analysis,2 pCi/g dose-equivalent ofI-131 is assumed as the activity concentration -
in the released coolant [23]. 'Monticello Technical Specifications allow either TID-14844 [24] or Regulatory Guide 1.109 [25] thyroid dose conversion factors (DCF) for' the definition of dose-equivalent I-131. Coolant activities calculated based on Reg. Guide 1.109 DCFs are higher than those. calculated with TID-14844 DCFs. Both activity concentrations are presented in Table 3.3-2.
' A portion of the released coolant exists as steam prior to blowdown, and as such does not contain the'same concentration per unit mass as the steam generated through blowdown. Therefore, it is necessary to separate the initial steam mass from the total mass released and assign a certain percentage (2% carryover .
fraction, in our analysis) of the fission product activity contained in this portion'of ' I steam by an equivalent mass of primary coolant. Table 3.3-3 shows the calculated equivalent mass in each hot standby case.
3.3.4 Offsite Dose and Control Room Dose Evaluations l- Activity released to the environment due to a main-steamline-break accident was i calculated based both on coolant concentrations with TID-14844 dose conversion l factors and coolant concentrations with Reg. Guide 1.109 dose conversion factors. )
Results of calculations are given in Table 3.3-4.
. ' Offsite dose consequences are presented in Table 3.3-5. .
Control room dose consequences are presented in Table 3.3-6. ,
l
. i l
22
GENE-B2100594-1 TABLE 3.3-1 ASSUMPTIONS FOR MONTICELLO RERATE MSLBA ANALYSIS PARAMETER VALUES REF Power Level (MWt) Hot standby at 4% rated [20]
power (66.8 MWt)
RPV Pressure (psia) [20]
Case 1 1158 Case 2 965-Time Elapse for MSIV Full Closure (sec) 10.5 (2)
Mass of Steam-Water Mixture Leaving Break (Ib) Table 3.3-3 [20]
Fuel Rod Damaged 0 (2)
Noble Gas Offgas Release Rate 300,000 C1/sec [2]
@ 30 min delay Noble Gas Offgas Release Fraction (21] -
Kr-83m 9.36E-03 i Kr-85m 1.64E-02 '{
Kr-85 ' 6.40E-05 ]
Kr-87 5.11 E-02 1 Kr-88 5.24E-02 !
Kr-89 2.18E-01 '
Xe-131m 5.23E-05 Xe 133m 7.82E-04 ,
Xe 133 2.19E Xe-135m 6.41 E-02 .
Xe-135 ' 5.92E-02 Xe-137 2.88E-01 Xe-138 2.18E-01 Normal Reactor Coolant Concentration (pCi/cc) [2]
l-131 4.06E-03 1-132 1.78E-02 1-133 1.50E-02 1-134 3.83E-02 {
l135 1.35E-02 i
j 23 l
GENE-B2100594-1 TABLE 3.3-1 (CONTINUED)
ASSUMPTIONS FOR MONTICELLO MSLBA DOSE ANALYSIS' (ARAMETER VALUES REF Data for Control Room Volume of Control Room (ft') 27,000 [15)
Filter Intake (cfm) 900 [17]
Efficiency of Charcoal adsorber (%) 98 [16]
Unfittered inteakage (T < 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) (cfm) 250 [16]
' Unfiltered inleakage (T > 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) (cfm) 10 [16] _
Recirculation Rate (cfm) 0 [2]
Ocbupancy Factor (18]
O - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.0 1 - 4 days 0.6 4 - 30 days 0.4 Control Room Intake Atmospheric [9]
Dispersion Factors (sec/m') 1 Ground Level Release O - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.67E-03 l 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.41 E-03 l 1 - 4 days 9.65E-04
! 4 - 30 days 5.62E-04 .
Offsite Atmospheric D!spersion Factor (sec/m') [9]
Ground Leve,I r
0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (EAB/LPZ) 9.20E-4/7.93E-5 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (LPZ) 7.93E-05 I 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 5.35E-05 1 - 4 days ,
2.28E-05 4 - 30 days 6.68E-06 l
1 24
___________D
GENE-B21005941 l
i i
TABLE 3.3-2 MSLB,A COOLANT CONCENTRATION - REDUCED TECH SPEC 2 pCi/g DE-131 TID 14844 RG 1.109 ISOTQff DCF (rem /Ci) Max cone (uCi/a) DCF (rem /Ci) Max cone (uCi/a)
, {
l-131 - 1.48E+06 0.77 1.49E+06 1.08 l-132 5.35E+04 3.38 - 1.43E+04 4.72 1-133 4.00E+05 2.85 2.69E+05 3.98 i.,34 2.50E.04 3.73e 03 10.,e 7.27 l
l-135 1.24E+05 2.56 5.60E+04 3.58 l
4 l
25 l
_ ____ _ _J
- p. -
GENE-B2100594-1 I i;
4 l
l TABLE 3.3-3 i MASS RELEASE FROM MSLBA Rorate Analyses USAR. Rev 13 l Case 1 Case 2 -
Powerlevel before accident (MWt) 1670 66.8 66.8.
Initial reactor pressure (psia) 1025 1158 965 Total mass released through break (ibm) 85000 86152 78617 Total steam released through break (Ibm) 20000 14578- 12394 Total liquid released through break (Ibm) 65000 71574 66223 Time for water level to cover steamline (sec) 2.00 -1.04 1.32 Initial steam released 7520 4030 4243 before steamline is covered (Ibm)
Equivalent liquid released from break (Ibm) 77630 82203 74459 i
l
)
l l l
}.
26 ,
GENE-B2100594-1 TABLE 3.3-4 MSLBA ACTIVITY RELEASE TO ENVIRONMENT (Cl)
TlD 14S44 DCF RG1.109 DCF ISOTOPE QASE1 CASE 2 CASE 1 CASE 2 1131 2.88E+01 2.62E+01 4.02E+01 3.64E+01 1132 1.26E+02 1.14E+02 1.76E+02 1.60E+02 1133 1.06E+02 9.63E+01 1.48E+02 1.35E+02 l134 2.71E+02 2.46E+02 3.79E+02 3.43E+02 1135 9.56E+01 8.66E+01 1.34E+02 1.21 E+02
{
TOTAllODINE 6.28E+02. 5.69E+02 8.77E+02 7.95E+02
]
KR-83m 1.33E 1.33E 01 1.33E-01 1.33E-01 KR-85m 2.32E-01 2.32E-01 2.32E-01 2.32E 01 I KR-85 9.07E 04 9.07E-04 9.07E-04 9.07E-04 KR-87 7.24E 01 7.24E-01 7.24E 01 7.24E-01 KR-88 7.43E-01 7.43E-01 7.43E-01 7.43E-01 KR-89 3.09E+00 3.09E+00 3.09E+00 3.09E+00 XE131m 7.41E-04 7.41E-04 7.41E-04 7.41E-04 XE133m 1.11E-02 1.11E 02 1.11E-02 ' 1.11E 02 XE-133 3.10E 01 3.10E-01 3.10E-01 3.10E 01 l XE135m 9.09E-01 9.09E-01 9.09E-01 9.09E 01
{
XE 135
~
8.39E-01 8.39E-01 8.39E-01 8.39E-01 XE 137 4.08E+00 4.08E+00 4.08E+00 4.08E+00 XE 138 3.09E+00 3.09E+00 3.09E+00 3.09E+00 TOTAL NOBLE GAS 1.42E+01 1.42E+01 1.42E+01 1.42E+01 i
i L
27 p
L_ _ _
l GENE-B21005941 l
l TABLE 3.3-5 MONTICELLO RERATE MSLBA OFFSITE DOSE (REM) 2-HOUR EA 30-DAY LPZ Thyroid Whole Body Thyroid Whole Body TfD DCF Case 1 17.3 0.28 1.49 0.02 Case 2 15.7 0.26 1.35 0.02 1
REG GUIDE DCF Case 1 24.2 0.40 2.08 0.03 Case 2 21.9 0.36 1.89 0.03 10CFR100 Limits 300 25 300 25 l i
i l
28 f
f
GENE-B2100594-1 TABLE 3.3-6 MONTICELLO RERATE MSLBA CONTROL ROOM DOSE (REM)
THYROID .WJ TID DCF Case 1 7.26 , 0.003 Case 2 6.58- 0.002 B_EG GUIDE DCF Case 1 10.1 0.004 Case 2 9.18 0.003 GDC19 Limits 30 5 l
l i
29 l
L E _
i
_____.-._,,.---c ,
- - ' = - " ~ - ' - - ' ' ~ ~ _ - _ - - _ - -- -- - - ~
, hn s
- o : GENE.B2100394-t :
g ,
m 3 ,
x 13.4 Fuel-Handlina Accident
"?
. L g.
l T The radiological evaluation of: fuel-handling accident l(FHA) 'is described,in LMonticello USAR Section 14.7.6 [2]. a-sf 13.4.1' Assumptions - y h . The fuel-handling accidents analyzed in the USAR assumed that the drywell head
. and the reactor vessel head are removed, and a fuel bundle was accidently dropped ;
on the core. The drop height into the fuel pool will be less than that into the com.
f Therefore a fuel bundle dropped on top of the core results in more damaged rods.
r.- . .
~ The number of rods assumed to fail in a fuel-handhng accident is dependent both -
- on the fuel design and the de's ign of fuel handling equipment. The power rerate; li "
y - core design is expected to be based on Gell or newer fuel designs.' Licensing analyses for gel I and GE13 fuels with 9x9 array and GE12 fuel with 10x10 array were recently reported in GESTARII [3]. The number of rods was calculated to
' L be 140 for 9x9 array' fuel and:172'for 10x10 array fuel. By comparison, the
- prc's ent analysis and the USAR both conservatively assumed failure of 125 rods ,
' with GE 8x8 fuel in their evaluations.- The relative amount of activities released' for 9x9 array fuel (74 full length fuel rods per htmdle) am (140/125)(60/74) = 0.91
. times'the activity released for a core of 8x8 fuel.' Similarly, the relative amount of '
activity released for 10x10 array fuel (87.33 full length' fuel rods per bundle) is'
~
(172/125)(60/87.33) = 0.95 times the activity micased for a core 'of 8x8 fuel.
Therefore, the present assumption boureds other fuel designs for fuel-handling l _ accident analysis.
The fuel handling accident analyzed in the USAR sssumed a drop height of less than L30 feet, resulting in 125 damaged rods.' GESTARll also assumed the k accident occured on top of the core. The analysis considered a drop height of 34 ll feet, resulting in 104. damaged fuel rods. Therefore, assuming 125 rods failed in y the present analysis provides extra conservatism in the evaluation.
( The accident parameters and assumptions are summarized below and listed in Table 3.4-1.
L 'T e The reactor has been operated at 102% rated po.wcr, or 1918 MWt, prior
. to refueling [1].
The accident occurs 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown (26].
- The GE generic fission products inventory at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay will be used to simulate thecore inventory at the time of accident [8].
e 125 rods in an 8x8 array fuel are assumed to be damaged [2].
30
g- __- . _ __ - - _ _ - _ - _ _ _ - - - _ - _ _ _ _ - _ _ _ _ - _ _ _ - _ _ - _ _ _ _ - _ _ __ . _ - - _ _ _ . _
GENE.B2100594 1 i
j' e' = lt is assumed that the damaged rods had power peaking factor 1.5 [26].
.i L e-It is assumed that the gap / plenum activity consists of 10% of noble gas y (except Kr-85, which is 30%) and 10% of iodine rod inventory [26].
e The effective fuel pool decon factor for all iodines is 100 [26].
- The SGTS filter efficiency is 85% for iodines [15].- I e : Activity released from fuel pool is airborne instantaneously and released to
- environment through SGTS with no hold up in the reactor building [17].
! - Regarding the last assumption, Regulatory Guide 1.25 [26] recommends the-accident evaluation to be based on activity release for two hours. Ifowever; in this analysis it is conservatively assumed that, regardless of limited flow rate through the SGTS, all airborne activity is released to the environment without delay.
Consequently, no credit is given for radioactive decay during activity releases. ;In the actual calculation, an arbitrarily large leak' rate (10' %/ day) is used to simulate the instantaneous exhaust of activity from the reactor building.
The GE generic fission products inventory at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown is presented in Table 3.4-2.
3.4.2 Results of Calculation
.Since all the airborne activity is assumed to be released instantaneously, j radioactivity in the environment is not a function of time. Calculated offsite' dose l
- - consequences are given in Table 3.4-3, together with 10CFR100 offsite osse ]
limits. )
i c
I l
1 I
J I
31 1
l I' Y . _ - _ - _ _ -_____ A
TABLE 3.4-1 ASSUMPTIONS FOR MONTICELLO RERATE FHA ANALYSIS PARAMETER VALUEJ REF Power Level (MWt) 1918 [1]
Fuel Rod Damaged (8x8 fuel) 125 [2]
Number of Fuel Bundles in Core 484 [2]
Number of Rods per Bundle 60 [4]
Core Inventory at Time of Accident Table 3.4-2 [8]
Power Peaking Factor for Damaged Rods 1.5 [26]
Percentage of Activity Released from Damaged Rods [26]
Halogen 10 Noble Gas 10 Kr-85 30 Pool Decontamination Factor for Halogens 100 [26]
Refueling Area Leak Rate (No hold up) 10' %/ day used in (17]
actual calculation Release Duration (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) irrelevant - all activity [25]
released instantaneously Secondary Containment Drawdown Time (sec) 0 [17]
SGTS Filter Efficiency (%) 85 [15]
Offsite Atmospheric Dispersion Factors (Sec/m ) [9]
Stack release 0 - 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (EAB/LPZ) 1.01 E-4/3.51 E-5 0.5 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (EAB/LPZ) 3.21 E-6/1.30E-6 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (LPZ) 1.30E-06 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 8.54 E-07 1 - 4 days 3.70E-07 4 - 30 days 1.11 E-07 32
GENE-B2100594-1 TABLE 3.4-2 CORE INVENTORY T=24 HR AFTER SHUTDOWN ISOTOPE Cl/MWt 1131 2.45E+04 l-132 3.16E+04 l133 2.53E+04 l-134 1.38E-03 1-135 4.15E+03 KR-83m 1.31E+01 KR-85m 1.66E+02 KR-85 3.02E+02 KR-87 2.58E-02 KR-88 4.82E+01 KR-89 6.25E-10 XE131m 1.58E+02 XE133m 2.09E+03 XE-133 5.33E+04 XE135m 6.36E+02 XE-135 1.22E+04 XE-137 1.35E-09 XE-138 1.28E-09 33
TABLE 3.4-3 MONTICELLO RERATE FHA OFFSITE DOSE (REM) 2-HOUR EA 30-DAY LPZ Thyroid Whole Body Thyroid Whole Body instantaneous release 2.04 0.192 0.71 0.067 10CFR100 Limits 300 25 300 25 1
l l
_ - _ _ _ _ _ _ _ _ _ _ _ _a
i-l; J p
i' j 4.'. REVIEW OF NRC COMMENTS l
u l Long term NRC Commitment M78027A [27] is not related to power rerate. It is l associated with the NSP 10CFR50.59 safety evaluation performed for the last fuel
- j. pool expansion.
For rerate, the fuel handling accident analysis is being submitted to the NRC for review and approval as part of the license amendment application. Section 3.4 of Lthis report shows that all NRC acceptance criteria continue to be satisfied under rerate conditions.
l l
I l
l l
l I
1
.o
, . . GENE B2100594-1
- b. <
by ,
- 5. REVIEW OF GENERIC COMMUNICATIONS '
o .,.
No ' generic. communications ' were found 'that were impacted by 'the Rerate Program for this topical area.-
t F',
l
, .L-3 I
I i
l l, .
I i
h ;
36
i i GENE-B2100594-1 i
e v
- 6. CONCLUSIONS ,
Design basis accidents considered' limiting for rerate operations are the control-
~" '
rod-drop accident, the loss-of-coolant accident, the main-steamline-break -
accident, and the fuel-handling accident. Radiological evaluation of these design basis accidents under rerate conditions concludes that' the regulatory limits of offsite doses at the exclusion area boundary and the low population zone are not exceeded. In addition,' control room doses due to these design basis accidents are -
within the GDC19 control room personnel exposure limits.
d
.j' .
37 u
___-___--__________________-_D
e GENE-B21 594-1 ,
l '..
)
l-7.-
SUMMARY
OF RESULTS.
u The7. design basis accidents considered inithis analysis are: control-rod-drop '
H -
accident, ' loss-of-coolant accident, main-steamline-break accident, and fuel-
- l. handling accident. Specific results' of each accident are presented in Tables 3.1-5,
- 3.2-3,3.3-5,3.3-6, and 3.4-3, respectively.
l .
I
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t l-i 8. RERATE EFFECTS ON RADIOLOGICAL CONSEQUENCES l
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. The power rerate radiological consequence analyses of design basis accidents are based on different sets of assumptions and source terms than those contained in the - -
- existing' USAR for rated power operations. The rerate analysis bases were revised in order to be consistent with current regulatory guidance.
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This section provides the dose consequence comparisons'for rated power (1670 MWt) versus rerated power (1880 MWt) under current' design basis assumptions, which 'are
- summarized in Tables 3.1 1, 3.2-1f 3.3-1, and 3.4-1 of this Report. In compliance
' with Reg; Guide 1.49, all the dose result corresponds to 102% of' operating power, or 1703 MWt and 1918 MWt.. respectively. Data representing rerate condition are extracted from accident analyses detailed in Section 3 of this report. Data representing rated power are ratioed from the rerated results by a power ratio of 1703/1918 (or 0.8879) and quoted in parentheses.
Three of the four design basis accidents in Monticello Rerate Radiological Analyses are included in the comparison described above. These are: Control-Rod-Drop Accident, Loss-of-Coolant Accident, and the Fuel-Handling Accident. ~ Each are l presented in Tables 8.1-1,8.1-2, and 8.1-4 respectively.
The radiological consequences from a main-steamline-break accident are proportional to' the mass of coolant released during the accident as' well as the activity concentration of the coolant. The current accident analysis is based on a different set of assumptions (hot standby) and source terms than those in the existing USAR, which assumed the accident occured at rated poiver operations. Under hot standby condition, the mass release is not adversely affected by power rerate [ Reference 20].
The Technical Specification coolant activity concentration is not affected by rerate either. Therefore, dose consequences of a main-steamline-break accident; at hot standby condition are independent of rated or rerated power (Table 8.1-3).
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GENE-B21005941 TABLE 8.1-1 MONTICELLO RERATE(RATED) CRDA OFFSITE DOSE (REM) 2-HOUR EA 30-DAY LPZ Thyroid ~ Whole Body Thyroid Whole Body Offsite Doses 0.92(0.82) 0.096(0.085) 0.66(0.59) 0.021(0.019)
SRP 15.4.9 Criteria (rem) 75 6 75 6 Note: Rerate data corresponds to 102% of 1880 MWt, or 1918 MWt.
Data in parentheses represent results at 102% of 1670 MWt, or 1703 MWt.
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GENE-B2100594-1 TABLE 8.1-2
- MONTICELLO RERATE (RATED) LOCA DOSE (REM)
CONTROL ROOM 2-HOUR EA 30-DAY LPZ Thyroid Whole Body Thyroid Whole Body Thyroid Whole Body 13.4(11.9) 0.020(0.018) 4.88(4.33) 0.83(0.74) 10.6(9.41) 0.36(0.32)
GDC19 Limits (rem) 10CFR100 Limits (rem) 30 5 300 25 300 25 Note: Rerate data corresponds to 102% of 1880 MWt, or 1918 MWt.
Data in parentheses represent results at 102% of 1670 MWt, or 1703 MWt.
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I GENE-B2100594-1 l
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TABLE 8.1-3 MONTICELLO RERATE (RATED)** MSLBA OFFSITE DOSE (REM) 2-HOUR EA 30-DAY LPZ Thyroid Whole Body Thyroid ,Whole Body TID DCF Case 1 17.3 0.28 1.49 0.02 Case 2 15.7 0.26 1.35 0.02 REG GUIDE DCF Case 1 24.2 0.40 2.08 0.03 Case 2 21.9 0.36 1.89 0.03 10CFR100 Limits 300 25 300 25 t-MSLBA CONTROL ROOM DOSE fREM)
TID DCF REG GUIDE DCF I i
Thyroid ySle Body Thyroid Whole Body Case 1 7.26 0.003 10.1 0.004 i
Case 2 6.58 0.002 9.18 0.003 GDC 19 Limits 30 5 30 5
" Note: Dose consequences correspond to hot standby condition, regardless of rated or rerated ;
operations. See text for detail.
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i GENE-B2100594-1 L
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TABLE 8.1-4 )
MONTICELLO RERATE(RATED) FHA OFFSITE DOSE (REM) 2-HOUR EA 30-DAY LPZ Thyroid Whole Body Thyroid Whole Body 2.04(1.81) 0.19(0.17) 0.71(0.63) 0.067(0.059) i 10CFR100 Limits 300 25 300 25 Note: Rerate data corresponds to 102% of 1880 MWt, or 1918 MWt.
Data in parentheses represent results at 102% of 1670 MWt, or 1703 MWt. l l
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GENE-B2100594-1 r.
- 9. REQUIRED ACTIONS L
The Monticello Rerate Design Basis Accident Analyses addressed in this report is
- a' complete revision from the Plant Safety' Analysis in the existing USAR. . The
, . : radioactivity source term as well as the analysis basis / assumptions all have been l- revised;to be' closely in compliance with current regulatory standards and
[. requirements. As a result, Chapter 14 of the Monticello Update Safety Analysis Report has to be revised for the Rerate Program.
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t GENE-B2100594-1 r
- 10. REFERENCES
- l. USAEC Regulatory Guide 1.49 Power Levels of Nuclear Power Plants, Rev.1, December 1973.
' 2. Monticello Updated Safety Analysis Report Rev. 13,1995.
- 3. NEDE-24011-P-A-11-US General Electric Standard Application for Reactor Fuel November 1995.
- 4. NEDE-31152P GE Fuel Bundle Designs Rev 4, February 1995.
- 5. NEDE-32417P GE12 Compliance with Amendment 22 of NEDE-240ll-P-A December 1994.
- 6. NUREG-0800 Standard Review Plan 15.4.9, Radiological Consequences of Control Rod Drop Accident (BWR) Appendix A Rev 2, USNRC 7/81.
' 7. US Regulatory Guide 1.77 Assumptions Used for Evaluating A Control Rod Ejection Accident for Pressurized Water Reactors, USAEC 5/1974.
- 8. GENE-B2100591-2 "Monticello Power Rerate Radiological Analysis Sources -
Task 21" Rev 0, February 1996.
- 9. Bechtel calculation 10040-UO30-M-01," Control Room and Offsite Atmospheric
- 10. US Regulatory Guide 1.3 Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors, USAEC 6/1974.
I1. Monticello Technical Specification.
- 12. NEDC-31858P BWROG Report for increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems Rev. 2, September 1988.
- 13. NUREG-0800 Standard Review Plan 15.6.5 Loss-of Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary, Rev.2, USNRC 7-1981.
- 14. NUREG-0800 Standard Review Plan 6.5.5 Pressure Suppression Pool As A l Fission Product Cleanup System, Rev. 0, USNRC 12/1988.
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GENE-B2100594-1
- 15. NSP Calculation CA96-094 " Standby Gas Treatment System Filter Efficiency and i
Operating Temperature under Rerate Conditions"
- 16. NSP Ictter Steve Hammer to P.T. Tran " Revised NSP comments on GLN-95-060, Radiological Analysis of Design Basis Accidents (Task 24)" l-9-96.
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- 17. NSP memo Dave Musolf to Steve Hammer "Information for Monticello Rerate DBA Analysis Data" 8-22-95.
- 18. NUREG-0800 Standard Review Plan 6.4 Control Room Habitability System, .
Rev. 2, USNRC 7/1981. i
- 19. NSP Monticello Power Rerate Task Report "High Energy Line Breaks System Engineering Evaluation - Task 27.0".
- 20. D. Pappone to S. Wang "Outside Steamline Break Mass Release for Monticello Power Rerate" I l-7-95.
- 21. General Electric proprietary version of Transient Reactor Analysis Code (TRAC)
NEDE-32176P TRACG Model Description Licensing Topical Report, 1:ebruary 1996.
- 22. NEDO-10871 Technical Derivation of BWR 1971 Design Basis Radioactive Material Source Terms, J. M. Skarpelos and R. S. Gilbert, March 1973.
- 23. Monticello Licensing Amendment Request Dated July 26,1996 " Reactor Coolant Equivalent Radioiodine Concentration and Control Room Habitability"
- 24. TID-14844 Calculation of Distance Factors for Power and Test Reactor Sites, ,
USAEC 3/1962. !
- 25. US Regulatory Guide 1.109 Calculation of Annual Doses to Man from l Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, USNRC 10/1977.
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- 26. US Regulatory Guide 1.25 Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for BWRs and PWRs,3/72.
L. 27. NRC Commitment Number M78027A Safety Evaluation for Amendment 34 Regarding Fuel Handling and Installation of the Modified Spent Fuel Rack, Completion date 4/18/78.
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Attachment 5 s-Annual Report Review of Meteorological Data 1991 .
Monticello and Prairie Island Stations l
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O Note to Attachment 5 l-This audit included a review of the deposition (D/Q) data. At the time of the audit, l the consultant could not duplicate the D/Q results and documented this l- >
discrepancy in the repott. The discrepancy was due to an inconsistency in the I
f consultant's calculation methodology and has since been resolved.-
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